This invention relates to a fuel assembly for a nuclear reactor and more particularly, to a fuel assembly for a boiling water reactor capable of ensuring a long operation cycle for fuel economy and of maintaining a high thermal margin.
A fuel assembly for a boiling water reactor (BWR) is constructed by a square channel box in which a number of fuel rods each comprising a metallic clad, fuel bundle, in which nuclear fuel material are packed are regularly arranged. The reactor core of the BWR includes a plurality of cells each comprising a cruciform control blade and four fuel assemblies surrounding the control blade, these cells being arranged in a regulated manner in the core. Namely, each fuel assembly and control blade have axes perpendicular and parallel to each other, and a coolant operated as a moderator flows from the lower portion towards the upper portion of the reactor core.
Concerning the BWR, steam voids are not formed in the bottom portion of the effective core portion, i.e. the lower end of a heat generating portion at which an exothermic reaction is performed, but many voids are generated at areas other than the bottom portion of the reactor core, and the void fraction in the moderator passage located above the central portion thereof is made considerably high, for example in excess of 70% near the upper end of the reactor core. As the void fraction increases, the flow speed of the coolant as moderator necessarily increases because the cross section in a direction normal to the axis of the coolant passage is made a constant height within the channel box. Friction of the passage increases in proportion to approximately the square of the flow speed of the coolant, so that the pressure loss rate of the coolant becomes large at the central portion of the reactor core toward the upper portion thereof.
The pressure loss rate of the coolant with respect to the unit length of the reactor core, .DELTA.P.sub.T /.DELTA.Z, will be represented by the sum of the following four factors. ##EQU1## in which .DELTA.P.sub.h : Position Loss,
.DELTA.P.sub.a : Acceleration Loss, PA1 .DELTA.P.sub.f : Friction Loss, and PA1 .DELTA.P.sub.L : Local Loss (due to the location of spacers, for example). PA1 A: Moderator Flow Passage Area, PA1 p: Moderator Density, PA1 g: Gravity Constant, PA1 f: Friction Loss Coefficient of Fuel Bundle and PA1 .DELTA.P.sub.f : Friction Loss of Fuel Bundle.
The most significant of these four factors in a position including no spacers is the friction loss, which is represented as follows: ##EQU2## in which M: Mass Flow Rate of Moderator (water),
The flow of the coolant is mainly caused by a drain pressure at the outlet port of a recirculation pump, and accordingly, the large pressure loss means that a large force has to be imparted to the pump, resulting in an enlargement of the machinery or system utilized and the consequent lowering of the power generation efficiency. The reduction of the pressure loss will therefore result in the reduction of the force or power to be imparted to the pump.
Recently, various studies or attempts have been carried out or tried in view of the improvement of the economics of the nuclear power plant. For example, it has been found out through these studies that the neutron multification factor during the reactor operation period is improved by changing the gaps between the respective fuel rods in the fuel bundle, thereby achieving the elongation of the operation cycle of the reactor and hence improving the burning efficiency. The improvement (increase) of the neutron multification factor during the reactor operation period is based on the improvement (increase) of the resonance escape probability due to the mutual shielding effects of the fuel rods, and on the fact that the extent of the disadvantage of the usage of the thermal neutrons, which may become somewhat disadvantageous, can be suppressed during the high-temperature reactor operation period. Japanese Patent Laid-Open Publication No. 75378/1987 of the same inventors and others as that of this application, discloses the above-described technique to some extent.
However, the fact that the gaps between the fuel rods are changed in the fuel bundle provides a problem regarding the configuration where there are portions through which the coolant passses easily and those for which it does not. This requires new means to be taken with respect to it. The smooth flow of the moderator will make it difficult to ensure a sufficient reactor thermal margin and the soundness of the fuel.